ML19205A520
| ML19205A520 | |
| Person / Time | |
|---|---|
| Site: | Seabrook  |
| Issue date: | 07/24/2019 |
| From: | Morgan, Morgan, Lewis & Bockius, LLP, NextEra Energy Seabrook |
| To: | Atomic Safety and Licensing Board Panel |
| SECY RAS | |
| References | |
| 50-443-LA-2, ASLBP 17-953-02-LA-BD01, RAS 55113 | |
| Download: ML19205A520 (5) | |
Text
UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION ATOMIC SAFETY AND LICENSING BOARD In the Matter of Docket No. 50-443-LA-2 NEXTERA ENERGY SEABROOK, LLC ASLBP No. 17-953-02-LA-BD01 (Seabrook Station, Unit 1)
Hearing Exhibit Exhibit Number: NER008 Exhibit
Title:
John Simons Curriculum Vitae
MPR Associates, Inc.
320 King Street Alexandria, VA 22314 703-519-0200 www.mpr.com John W. Simons EXPERIENCE
SUMMARY
Mr. Simons joined MPR in 1987 and his work has focused on nuclear power plants. He has directed test programs to assess degradation of concrete due to chemical processes. He has overseen MPRs efforts on license renewal and subsequent license renewal related to concrete degradation.
Mr. Simons is currently the General Manager of Projects, with responsibility for overseeing execution of all MPR projects (Power Services, Federal Services and Product Development). He previously served as Director Plant Systems & Components and General Manager of Power Services.
ACCOMPLISHMENTS
SUMMARY
Alkali-Silica Reaction in Concrete Boric Acid Attack of Concrete Directed multi-year program to evaluate the impact Directed effort to evaluate degradation of a Fuel of ASR on reinforced concrete structures at a Handling Building, a reinforced concrete structure, nuclear power facility. from prolonged exposure to boric acid from the Spent Fuel Pool. This effort included: multiple x The initial phase included: (1) extensive site testing programs to quantify the degradation rate of walkdowns to document the extent of concrete and embedded steel to boric acid; condition; (2) a limited-scope test program to analytical modeling of reaction kinetics; and assess the performance of post-installed assessment of the structural adequacy of the anchors in ASR-affected concrete; (3) building based on the postulated extent of assessment of the impact of ASR on the degradation. The effort demonstrated that the adequacy of plant structures and anchors. building will be structurally adequate through the x The second phase included execution of large- end of plant life, including consideration of license scale test programs to assess: (1) the effect of renewal. The effort also led to development of an ASR on shear strength, flexure, and aging management strategy to support license reinforcement anchorage using beams without renewal.
transverse reinforcement; (2) the impact of Oversaw efforts that developed multiple EPRI ASR on anchor capacity; and (3) evaluation of reports that provide guidance for evaluating and instrumentation for monitoring expansion. managing the aging effects associated with x The final phase included development of a leakage of borated water from PWR spent fuel license amendment request for a change in pools.
methodology for addressing ASR and support LOCA Thermal-Hydraulic Testing for enhancing the Structures Monitoring Provided technical support for the International Program and Aging Management Programs. Program on the Thermal-Hydraulic Behavior of Played a key role in interfacing with the regulator ECC during the Refill and Reflood Phases of a throughout the 8-year effort. Developed a LOCA in a PWR (2D/3D Program). This effort presentation for an all-day ACRS meeting on included:
management of ASR. x Specification and evaluation of multiple test Provided oversight of efforts for EPRI to: series at the full-scale Upper Plenum Test Facility (UPTF); and x Develop a screening approach for nuclear plants to determine their susceptibility to ASR, x Preparation of reports which analyzed the test investigate its presence and determine the results, compared results to subscale tests extent of condition. and TRAC simulations, and assessed their implications for reactor safety issues.
x Develop generic aging management guidance.
Served as principal editor of summary reports covering 15 years of research.
John W. Simons Page 2 of 4 Engineering Leadership of Large Projects Extended Power Uprates Provided engineering leadership and management Involved with MPRs efforts related to Extended on large projects at nuclear power plants. MPR Power Uprates (EPUs), including directing several developed the procurement specifications, of the efforts. These efforts include the following.
evaluated bids, and provided the primary interface with vendors on technical and schedule issues. x Independent assessment of scoping study, MPR provided oversight of design change package project cost estimates and project plan for the preparation by the architect engineer, and resolved proposed EPU to support the board of technical issues during installation. directors decision to approve the EPU.
x Consulting on key decisions related to an EPU x Feedwater Heater and MSR Tube Bundle including: selection of the thermal power Replacement. Acted as Project Engineer for target, evaluation of options for implementing replacement of six low pressure FWHs and six an EPU given uncertainty in the schedule for moisture separator reheater tube bundles. He obtaining the necessary regulatory approvals, was the primary interface with the vendor on and the decision to defer EPU implementation.
all technical issues and a key contributor to successful negotiations with the vendor on x Engineering support for component back charges. repairs/upgrades to support an EPU:
moisture-separator upgrades; and FWH shell x Nuclear Power Plant Security Upgrades. repairs and re-rating.
Directed MPR effort to provide project engineering and construction management x Extent-of-Condition review to assess EPU-services for a permanent checkpoint facility to related vulnerabilities following the steam dryer satisfy the revised Design Basis Threat. The issues.
upgrades were completed by the NRCs x Independent review of vendor studies that deadline and well-received by NRC. Also, evaluated the impact of the BWR and PWR acted as Technical Lead for upgrades to power uprates on plant systems and ensure success in a graded Force-on-Force components.
exercise. These upgrades were instrumental in decisive victories in the graded Force-on- EPRI MOV Performance Prediction Program Force exercise. Coordinated preparation of the EPRI MOV Performance Prediction Program Topical Report, x Generator Step Up Transformer which summarizes the entire program and presents Replacement. Managed MPR effort to the thrust and torque prediction methods provide project engineering support for developed in the program. It also documents the replacement of a generator step up (GSU) technical justification for the program and transformer and construction of a storage describes the individual testing efforts performed facility for spare GSU transformers. The within the program. This report is the official transformer replacement was successfully licensing submittal of the Program and constitutes completed during a 20-day outage. the basis for the NRC's review of the EPRI CFCU Modifications Program.
Managed MPR project to simplify the portion of the Pressure Locking/Thermal Binding of MOVs Service Water system that supplies Containment Developed and applied an innovative method for Fan Cooler Units at a nuclear plant. The project predicting bonnet pressurization due to valve heat reduced the required flow rate during accident up which accounts for expansion of the bonnet with conditions and adopted a single-flow for normal pressure and temperature. Predictions of the and accident conditions. The project also pressure increase agree favorably with test data, addressed silt accumulation in the system. The and are markedly less than would be predicted MPR effort included: conceptual design for the assuming a rigid body. This model has been modifications; determination of the new flow to the applied to valves at several nuclear sites.
CFCUs during accident conditions; all thermal-hydraulic analyses; and support for a license Performed transient heat transfer calculations to amendment request. The simplified system determine the change in bonnet fluid temperature significantly reduced operator burden and system due to ambient temperature transients or system maintenance requirements. temperature transients. Results of these calculations were input into calculations of bonnet pressurization using the model described above.
John W. Simons Page 3 of 4 Feedwater Heater Shell Inspection & Repair Feedwater Heater Support Directed MPR efforts related to feedwater heater Provided engineering support for FWH-related (FWH) shell thinning. problems at coal-fired and nuclear power plants.
Selected tasks are identified below.
x Provided engineering support for repair of FWH shells to address thinning in the vicinity x Post-mortem on Failed Heaters. These of the steam inlet nozzles. Repairs restored efforts included inspection of disassembled the heater shell to compliance with the ASME heaters and on-site supervision of heater Code and modified heater internals to mitigate disassembly. Recommendations for how to future thinning. avoid similar failures in other heaters were provided.
x Developed minimum wall thickness criteria for interim operation of FWHs with thinned shells. x Heater Replacement Evaluations. Assessed The criteria maintain ASME Code margins for current condition of heaters, projected future internal and external pressure for general and condition, and established schedule for heater localized thinning. replacements. This work also involved identification of options for addressing x Developed an improved replacement shell continued degradation of the heaters and section design (clad shell section with clad assessing the impact of the options on steam nozzle) to provide a permanent solution feedwater system operation and overall plant to shell thinning at the steam inlet nozzle. capacity.
Provided engineering support for fabrication and installation of the replacement shell x Nondestructive Examination (NDE).
sections. Provided engineering support for NDE of tubes in high-pressure FWHs at supercritical fossil x Managed MPR project for engineering support stations. This work involved determination of for re-rate and repair of FWH shells. This NDE scope, interpretation of NDE results, and project included: design/specification of recommendation of tubes to be plugged. This replacement shell sections and external shell task reduced loss-of-generation due to FWH stiffeners, engineering support during tube leaks by identifying seriously damaged fabrication, ASME Code design report for the tubes and plugging them before they fail in re-rate/repairs, definition of code compliance service.
approach (ASME Code, NBIC, state regulations), contingency planning, consulting x Insurance Claims for Feedwater Heaters.
on DCP development, and on-site support Supported plant in successful insurance claims during installation. The project covered 24 for feedwater heaters. This effort included FWHs (36 steam inlet nozzles). development of the technical basis for the claim, and quantification of the monetary value Circulating Water Intake Improvements of the damages.
Directed MPRs efforts to evaluate alternatives for seasonal reduction in Circulating Water System Tritium Leakage Task Force flow to reduce ingress of detritus into the plant. Participated on a task force at a commercial The MPR scope included evaluation of nuclear power plant to determine the source of alternatives, quantification of the improvement in tritium leakage from high concentration plant net electrical output of the plant, and development sources to the external environment.
of a conceptual design for using variable frequency drives. Buried Pipe Inspections Directed MPR efforts in support of inspection of Participated in the definition of an overall integrated buried prestressed concrete cylinder pipe (PCCP) strategy for improvements to address grassing by at a nuclear power plant. These efforts included reducing flow, replacing traveling and relocating (1) development of acceptance criteria for the the detritus return. This included scoring of inspections, (2) analyses to demonstrate that the different combinations of options to determine the as-found condition of flawed bell-and-spigot joints best approach. was suitable for continued service, and (3) development of contingency repair designs.
John W. Simons Page 4 of 4 EDUCATION University of Maryland, B.S., Chemical Engineering, 1987 Columbia Union College, B.S., Chemistry, 1986 PUBLICATIONS EPRI Reports EPRI MOV Performance Prediction Program: Topical Report. EPRI, Palo Alto, CA: 1994. TR-103237.
Program on Technology Innovation: Approach to Transition Nuclear Power Plants to Flexible Power Operations. EPRI, Palo Alto, CA: 2014. 3002002612.
Aging Management for Leaking Spent Fuel Pools. EPRI, Palo Alto, CA: 2016. 3002007348.
Welding and Repair Technology Center: Boric Acid Attack of Concrete and Reinforcing Steel in PWR Fuel Handling Buildings. EPRI, Palo Alto, CA: 2012. 1025166.
Best Practices for Evaluation of Spent Fuel Pool Leakage in Pressurized Water Reactors, included as an appendix to: Advanced Electromagnetic Inspection Methods for Fuel Pool and Transfer Canal Liners. EPRI, Palo Alto, CA: 2012. 1025214.
Tools for Early Detection of ASR in Concrete Structures. EPRI, Palo Alto, CA: 2015. 3002005389.
Papers and Presentations White, P., S. Bernhoft, A. Roberts, J. Simons, An Approach for Migrating Nuclear Power Plants to Flexible Power Operations, Proceedings of the 9th Nuclear Plants Current Issues Symposium, pp. VII 3/1 to VII 3/14.
Simons, J., R. Noble, C. Bagley, J. Moroney and R. Vayda, Approach for Large-Scale Testing of Concrete Affected by Alkali-Silica Reaction, Proceedings of the 9th Nuclear Plants Current Issues Symposium, pp. VIII-3/1 to VIII-3/21.
Noble, R. and J. Simons, ASR in Reinforced Concrete Structures: A Comprehensive Program to Address ASR at Seabrook, EPRI Symposium on Concrete Structures, Liner Barriers and Tanks, May 6 8, 2014.
Simons, J., C. Bagley and R. Vayda, Nuclear Power Plant Concrete Aging Management Issues, Proceedings of the 8th Nuclear Plants Current Issues Symposium, pp. VII-4/1 to VII-4/20.
Antoniazzi, A. and J. Simons, Case Studies on Tritium Pathways, EPRI Buried Pipe Integrity Group Meeting, February 15-17, 2011.
Simons, J., Generic Approach for Assessing Impact of SFP Leakage on Reinforced Concrete Structures, 2nd EPRI Fuel Pool, Transfer Canal, and Concrete Degradation Issues Workshop, May 11-13, 2010.
Simons, J. and R. Keating, Managing Feedwater Heater Shell Thinning, 10th International Conference on Nuclear Engineering, April 2002.
Bradley, M., J. Simons, D. Strawson and D. Chapin, Overview of Dry Storage of Spent Nuclear Fuel in the U.S., Seventh International Conference on Nuclear Engineering, April 1999.
Claude, E.J., H. Nakamura, D.M. Chapin, J.W. Simons and H. Seneviratne, Update on Decommissioning Waste from Nuclear Power Plants, 7th International Conference on Nuclear Engineering, April 1999.
Damerell, P.S., D.H. Harrison, P.W. Hayes, J.W. Simons, and T.A. Walker, "Effects of Pressure and Temperature on Gate Valve Unwedging," Fourth NRC/ASME Symposium on Pump and Valve Testing, NUREG/CP-0152, pp. 3C-49 to 3C-71.
Rhee, G., P. Damerell and J. Simons, "Use of 2D/3D Data to Scale Up Liquid Carryover/De-entrainment (Steam Binding) Phenomena to a PWR," 16th Water Reactor Safety Information Meeting, October 24-27, 1988.
NUREGs Damerell, P.S., and J.W. Simons, editors, 2D/3D Program Work Summary Report, edited by MPR Associates, NUREG/IA-0126, June, 1993.
Damerell, P.S., and J.W. Simons, editors, Reactor Safety Issues Resolved by the 2D/3D Program, edited by MPR Associates, NUREG/IA-0127, July, 1993.